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PENGARUH ROL DINGIN TERHADAP SIFAT MEKANIK MATERIAL AL 6061-T6 Saga Octadamailah; Dwi Marta Nurjaya; Supardjo; Ganisa Kurniati Suryaman
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 1 (2021): Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2021.27.1.6339

PENGARUH PROSES PENGEROLAN DINGIN TERHADAP SIFAT MEKANIK MATERIAL AL 6061-T6. Proses pembuatan pelat tutup dan pelat bingkai untuk pelat elemen bakar reaktor G.A. Siwabessy dilakukan dengan cara dirol dingin, sehingga dicapai ketebalan yang diinginkan. Pada penelitian sebelumnya, proses pengerolan dingin pada material AlMg2 menyebabkan perubahan sifat mekanik, dimana material menjadi lebih ulet dan kekuatan tariknya menurun. Pada bahan bakar densitas tinggi, material kelongsong Al 6061-T6 akan digunakan untuk menggantikan material kelongsong sebelumnya, yaitu AlMg2. Tujuan dari penelitian ini adalah untuk mengetahui perubahan sifat mekanik dari material Al 6061-T6 hasil pengerolan. Untuk melihat perubahan sifat mekanik yang terjadi, dilakukan uji tarik, pengamatan mikrostruktur dan struktur kristal. Hasil uji tarik menunjukkan bahwa kekuatan tarik, modulus elastis dan kekuatan luluh cenderung naik seiring dengan besarnya derajat reduksi ketebalan. Meskipun demikian, kekuatan luluh dari hasil pengujian cukup rendah. Rasio antara kekuatan luluh dan kekuatan tariknya hanya 0,7. Dari hasil pemindaian menggunakan difraksi sinar-X diketahui bahwa proses pengerolan dapat menyebabkan penghalusan ukuran kristalit. Kesimpulan dari penelitian ini secara umum adalah proses pengerolan dapat meningkatkan sifat mekanik material Al 6061-T6.

PERCOBAAN PEMBUATAN PELAT ELEMEN BAKAR U-7Mo-xSi/Al DALAM UKURAN MINI Supardjo; Agoeng Kadarjono; Isfandi; Yatno Dwi Agus Susanto; Setia Permana; Guswardani
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 25 No. 3 (2019): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2019.25.3.5697

The purpose of this study is to obtain the optimized process parameters in the fabrication of a U-7Mo-xSi/Al mini plate complying with the specifications. The process flow started with synthesizing U-7Mo-xSi alloys (x = 1%, 2%, and 3%) using depleted uranium and continued to preparing U-7Mo-xSi powder, U-7Mo-xSi/Al fuel core, and U-7Mo-xSi/Al fuel element plates. The quality of the product was determined by analyses/tests including tests of raw materials and products from each end of the process by the appropriate type of test which is commonly applied to the manufacturing process of plate type fuel. The quality of U-7Mo-xSi alloy ingots, U-7Mo-xSi powder, and U-7Mo-xSi/Al fuel core was satisfactorily good and in compliance with the specifications, indicating that the applied process parameters were appropriate. The obtained fuel element plates, however, have not satisfied the requirements especially the thickness of the cladding, where there are still several measurement points whose thickness is <0.25 mm. To obtain cladding thickness according to the required specifications, it is necessary to experiment with other rolling parameters or replacing the cladding material with material of higher hardness.

PRODUCTION OF URANIUM-MOLYBDENUM ALLOY AS A CANDIDATE FOR NUCLEAR RESEARCH REACTOR FUEL Ganisa Kurniati Suryaman; Muhammad Waziz Wildan; Supardjo; Yatno Dwi Agus Susanto
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 3 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2018.24.3.4685

Research and development on high density uranium for nuclear research reactor fuel is still in progress. Uranium-molybdenum alloy is one of the strongest candidates of nuclear research reactor fuel material. The properties and characteristics of U-Mo alloy is of important consideration for the selction of the fabrication techniques for the production of the fuel. In this work, uranium-molybdenum (U-Mo) alloys with varied molybdenum content have been produced succesfully by arc melting technique. The molybdenum content variations were 7 %wt, 8 %wt, 9 %wt and 10 %wt Mo. The melting process was done 5 times to achieve homogenization. Metallographic micrograph shows the presence of dendritic structure. XRD examination result affirms the presence of 2 phases of γ-U phase and d-U2Mo phase. Microhardness Vickers test shows higher hardness value for Uranium-molybdenum alloy with higher molybdenum content.

KARAKTERISASI PADUAN AlMgSi UNTUK KELONGSONG BAHAN BAKAR U3Si2/Al DENSITAS URANIUM 5,2 gU/cm3 Aslina Br. Ginting; Supardjo; Yanlinastuti Yanlinastuti; Dian Anggraini; Boybul Boybul
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 24 No. 1 (2018): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2018.24.1.4016

Increase in uranium density from 2.96 gU/cm3 to 5.2 gU/cm3 of U3Si2/Al fuel requires a compatible cladding. High density fuel possesses high hardness value, and the use of AlMg2 as cladding material may cause dog bone during rolling process. In addition to dog bone phenomena, swelling can also occur during the use of nuclear fuel in the reactor due to rising in fission products and burn up. For this reason, substitution of cladding material should be determined to suit the high density U3Si2/Al fuel. This research deals with the characterization of AlMgSi as a candidate for AlMg2 substitution. The characterization includes thermal analysis with DTA (Differential Thermal Analysis) dan DSC (Differential Scanning Calorimetry), hardness test with macrohardness tester, microstructure examination with SEM (Scanning Electron Microscope), and corrosion analysis by heating at 150 °C for 77 hours in an autoclave. Analysis results show that both AlMgSi and AlMg2 share stable heat compatibility with U3Si2/Al up to 650 °C. AlMgSi cladding has a hardness value of 115 HVN, higher than that of AlMg2, which is 70.1 HVN. Microstructure examination shows that AlMgSi has better morphology of interface bonding than AlMg2. Corrosion rate analysis also indicates that AlMgSi has lower corrosion rate. The results of thermal, hardness, microstructure and corrosion characterization reveal that the U3Si2/Al fuel of 5.2 gU/cm3 suits better with AlMgSi than AlMg2.

OPTIMASI PARAMETER METODE PEMISAHAN RADIONUKLIDA 95Zr DALAM LARUTAN PELAT ELEMEN BAKAR U3Si2/Al TERIRADIASI Yanlinastuti Yanlinastuti; Noviarty Noviarty; Iis Haryati; Ariyanti Saputri; Boybul Boybul; Supardjo
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 3 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2021.27.3.6535

Radionuclide of 95Zr is one of the γ-radiation emitter radionuclides which has a short half-life so that it can be used as an indicator of burn up. In the irradiated U3Si2/Al fuel element plate there are several radionuclides of fission products emitting α, β and γ and when measuring radiation using γ-spectrometry, these radionuclides may interfere with each other resulting in inaccurate measurement results. Therefore, it is necessary to separate 95Zr for better accuracy in the measurement of γ-radiation with a γ-spectrometer. The purpose of this study was to obtain the optimum parameters for the separation of 95Zr in U3Si2/Al fuel element plate solution by ion exchange column method using Dowex 1x-8Cl- resin of 100-200 mesh in diameter. The optimum parameters obtained were then used for the separation of 95Zr in a solution of irradiated U3Si2/Al fuel element. It was found that the optimum parameter of acidity of H2SO4 in the feed was a concentration of 0.5 M, and the acidity of H2SO4 for elution was 1.0 M with a recovery of 98.20%. The weight of the resin to bind 95Zr radionuclide was 2.0 g with a recovery of 78.76%. The required feed volume was 50 µL with a recovery of 96.90% and the optimized flow rate was 0.1 mL/minute with a recovery of 96.72%. The optimum values were used for the separation of the U3Si2/Al fuel element plate solution irradiated with the code of Bottom (B), Middle (M) and Top (T). The measurement results show that the mean recovery of separation was 95Zr for code B-1 = 75.078%; B-2= 81.401%; M1=76,850%; M-2=83.806%; T-1=84.433%; and T-2 = 81.728% with the acceptability of the CV repeatability value smaller than the CV Horwitz value.

PENGARUH DENSITAS URANIUM TERHADAP UMUR BAHAN BAKAR NUKLIR DI DALAM REAKTOR RSG-GAS DITINJAU DARI ASPEK NEUTRONIK Saga Octadamailah; Supardjo
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 23 No. 2 (2017): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2017.23.2.3550

Material Testing Reactor (MTR) is a research reactor with characteristics: small power, high thermal neutron flux and short life time fuel. Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) located in Serpong is an MTR type reactor initially operated with dispersion fuel of U3O8/Al with uranium enrichment of 19.75 % 235U and uranium density of 2.96 gU/cm3. In an effort to increase the lifetime of nuclear fuel, BATAN has done research on high density uranium fuel starting from U3Si2/Al fule and continued with UMo based fuel in line with investigations performed by worldwide researchers in the field of research reactors. Fuel of U3Si2/Al with a density of 2.96 gU/cm3 has been successfully manufactured and utilized as fuel for RSG-GAS to substitute the U3O8/Al fuel. Post irradiation examination has also been done for U3Si2/Al with a density of 4.8 gU/cm3 up to a burn up of 60 %, while research on UMo with 7 gU/cm3 density has been progressing into mini plate fabrication. Research on high density fuel has been focused on fabrication process, while little has been done on lifetime calculation or utilization period. Therefore, this research has performed a calculation on nuclear fuel lifetime using a paired program of ORIGEN and MCNP. The calculation results show that an increase in uranium density will extend the lifetime of fuel exponentially. The longest lifetime is achieved by UMo type, which extends to 651 days, and this relates to a lifetime extension of 219 days if the RSG-GAS is operated with UMo fuel.

KOMPOSISI, STRUKTUR DAN DENSITAS PADUAN U-7Mo-xSi Supardjo; Agoeng Kadarjono; Jan Setiawan; Ganisa Kurniati Suryaman
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2021.27.2.6386

The structure of U-7Mo-xSi (x = 1, 2, and 3%) alloy was studied to obtain a new uranium alloy suitable for use as a candidate for high-density uranium research reactor fuel. The U-7Mo-xSi alloy was made by melting technique using an arc furnace. The characterization of the alloy includes chemical composition, density, and structural phases to identify the phases contained in the alloy. The hardness test was also done. The density characterization results show that U content and density of the alloy decrease with increasing of Si content in the U-7Mo-xSi alloy. On the other hand, the hardness of the alloy increases with increasing of Si content. The X-ray diffraction pattern of U-7Mo-xSi alloy was observed from diffraction angle of 25° to 95°. The results indicate the presence of U-γ, U3Si2, and U3Si2Mo phases. The higher the Si content, the more U3Si2 phase is formed and the more the hardness increases due to the formation of bonds between Si and Si.

KARAKTERISASI TERMAL PEB U-7Mo/Al SEBAGAI BAHAN BAKAR REAKTOR RISET Sayyidatun Nisa; Ariyanti Saputri; Sutri Indaryati; Noviarty Noviarty; Supardjo; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 27 No. 2 (2021): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2021.27.2.6420

UMo/Al-based fuel is a candidate for research reactor fuel and as an alternative to U3Si2/Al fuel. This is because the UMo/Al alloy has a higher uranium density of 16.4 g/cm3 compared to U3Si2/Al of 12.2 g/cm3. In addition, the UMo alloy has a low neutron absorption cross section and its reprocessing process is relatively easy. The U-7Mo/Al alloy is able to maintain the γ-U phase during irradiation process so that the stability of the fuel in the reactor is maintained. The U-7Mo alloy has been fabricated into a U-7Mo/Al mini fuel element plate. The purpose of this study was to determine the thermal characteristics of U-7Mo/Al as inputs to the research reactor fuel fabricator in the manufacture of UMo/Al-based fuel. Thermal characterization was carried out for uranium, molybdenum, UMo and the fuel element plate of U-7Mo/Al using Differential Scanning Calorimetry (DSC). The thermal characterization results show that the U-7Mo/Al fuel element plate is thermally stable up to 600 °C because it undergoes an endothermic thermochemical reaction at a temperature of 632.962 °C with DH = 144.318 J/g. At 656.186 °C, the Al matrix and AlMg2 cladding are fused in the U-7Mo/Al with DH = 144.318 J/g. At temperatures of 952.810 °C to 1106.050 °C, the molten Al matrix diffuses with UMo to form a metastable U(Mo,Al)X compound and heating to a temperature of 1211.760 °C forms a UAlX layer (UAl2, UAl3 and UAl4) with DH = –22,721 J/g.

PENGARUH TEMPERATUR DAN IRADIASI TERHADAP INTERDIFUSI PARTIKEL BAHAN BAKAR JENIS U−7Mo/Al Maman Kartaman Ajiriyanto; Aslina Br. Ginting; Supardjo; Boybul Boybul
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 1 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2016.22.1.2740

U−7Mo/Al alloy had great potential as research reactor fuel, but it had several disadvantages, such as, it can form a interaction layer at the interface during the process of fabrication and irradiation in a reactor. The research objective was determining the interaction layer that was caused by interdiffusion or diffusion couple of U−7Mo with AlMg2 alloy which was annealed at 500 °C and 550 °C for 24 hours. The observation of microstructure used a Scanning Electron Microscope (SEM) on diffusion couple sample which was heated at temperature of 500 °C had not seen the layer interaction or the formation of a new phase between particles of U−Mo and Al matrix, but heating treatment at 550 °C indicated that U−7Mo alloy had been interdiffusion with AlMg2 plate produced (U,Mo) Alx compound on the interfaces. It was evidenced by interdiffusion reaction analysis used DTA that showed that U−7Mo / Al alloy at 500 °C had good heat compatibility, but at temperatures upper than 550 °C it had been phase changed from a + d to a + g phase. The heating in DTA furnace up to 679.14 °C produced U(Al,Mo)x meta stable phase and then interdiffusion process with uranium molten formed layer interaction that formed UAlx compound agglomerates (UAl4, UAl3 and UAl2). Agglomerates was formed from the heating process which was similar to agglomerates that caused by irradiation. U−7Mo / Al Fuel alloy that had 58% burn up had been interdiffusion between U−7Mo with Al matrix produced U(Al,Mo)x metastable phase that turned into (U, Mo) Al7 layer, UMo2Al20 precipitates, (UMo)Al3−Al and formed a boundary or UAlx (UAl4, UAl3 and UAl2) agglomerates. The results of microstructure analysis used SEM and interdiffusion reactions used DTA was supported by the analysis of micro hardness used Vickers Hardness. The results of hardness analysis that was done to AlMg cladding and U−7Mo alloy (before and after heating) and diffusion couple of U−7Mo / Al samples with AlMg2 plate after heating at 550 °C were respectively 64.62 and 340.45 HV (before heating) and 52.34; 303.16 and 497.34 HV (after heating). Diffusion couple U−7Mo/Al with AlMg2 plate samples had the highest hardness value. This hardness difference showed that the interdiffusion test used diffusion couple produced a new compound (U, Mo) Alx in interface zone that had different character, but the formation of interaction layer is not expected in the fuel U−Mo / Al dispersion because micro hardness and density of (U, Mo) Alx compound’s layer was lower than the average density of U−7Mo/ Al alloy.

PENGARUH UNSUR Zr TERHADAP PERUBAHAN SIFAT TERMAL BAHAN BAKAR DISPERSI U-7Mo-xZr/Al Supardjo; Agoeng Kadarjono; Boybul Boybul; Aslina Br. Ginting
Urania: Jurnal Ilmiah Daur Bahan Bakar Nuklir Vol. 22 No. 1 (2016): Urania : Jurnal Ilmiah Daur Bahan Bakar Nuklir
Publisher : Badan Riset dan Inovasi Nasional

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/urania.2016.22.1.2741

Thermal properties data of nuclear fuel is required as input data to predict material properties change phenomenon during the fabrication process and irradiated in a nuclear reactor. Study the influence of Zr element in the U-7Mo-xZr/Al (x = 1%, 2% and 3%) fuel dispersion to changes in the thermal properties at various temperatures have been stiffened. Thermal analysis includes determining the melting temperature, enthalpy, and phase changes made using Differential Thermal Analysis (DTA) in the temperature range between 30 °C up to 1400 °C, while the heat capacity of U-7Mo-xZr alloy and U-7Mo-xZr/Al dispersion fuel using Differential Scanning Calorimeter (DSC) at room temperature up to 450 °C. Thermal analyst data DTA shows that Zr levels of all three compositions showed a similar phenomenon. At temperatures between 565.60 °C – 584.98 °C change becomes α + δ to α + γ phase and at 649.22 °C – 650.13 °C happen smelting Al matrix Occur followed by a reaction between Al matrix with U-7Mo-xZr on 670.38 °C – 673.38 °C form U (Al, Mo)x Zr. Furthermore a phase change α + β becomes β + γ Occurs at temperatures 762.08 °C – 776.33 °C and diffusion between the matrix by U-7Mo-xZr/Al on 853.55 °C – 875.20 °C. Every phenomenon that Occurs, entalphy posed a relative stable. Consolidation of uranium Occur in 1052.42 °C – 1104.99 °C and decomposition reaction of U (Al, Mo)x and U (Al, Zr)x becomes (UAl4, UAl3, UAl2), U-Mo, and UZr on 1328,34 °C - 1332,06 °C , The existence of Zr in U-Mo alloy increases the heat capacity of the U-7Mo-xZr/Al, dispersion fuel and the higher heat capacity of Zr levels increased due to interactions between the atoms of Zr with Al matrix so that the heat absorbed by the fuel increase.

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